ML20087F328

Responds to RAI Re Proposed Conversion of Units 1 & 2 TS Based on NUREG-1431

ML20087F328
Person / Time
Site:	Vogtle
Issue date:	08/09/1995
From:	Mccoy C GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1431 LCV-0653, LCV-653, TAC-M92131, TAC-M92132, NUDOCS 9508150229
Download: ML20087F328 (55)

Text

/- peg a Power Company 40 lovemess Center Parkwaf Pv,1 Offce Box 1295 Bamengham, Alabama 35201 Teiephone 205 877 7122 L

C.K.McCoy Georgia Power  ;

&&$ """' August 9, 1995 ne saven ewc snem Docket Nos. 50-424 LCV-0653 50-425 TAC- M92131 M92132 U. S. Nuclear Regulatory Commission NITN: Document Control Desk Washington. D. C. 20555 Gentlemen:

i VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING Tile PROPOSED CONVERSION OF THE UNIT 1 AND UNIT 2 TECHNICAL EPECIFICATIONS BASED ON NUREG-1431 By letter dated May 1,1995, Georgia Power Company (GPC) proposed to amend the Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS), in part, by converting the existing VEGP TS to the format and content of NUREG-1431, the improved Standard Technical Specifications for Westinghouse plants. Among other things, the proposed conversion involves the relocation of existing VEGP TS requirements to a licensee controlled document. The relocation of these existing VEGP TS requirements is based on the criteria for the content of the TS set forth in the NRC Final Policy Statement on Technical Specification Improvement (58FR39132), July 22,1993.

Our May 1,1995, submittal referenced a May 1988 NRC letter from T. E. Murley to W. S. Wilgus, "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," for those items that were generically applicable to VEGP. In lieu of the reference to the above letter, the NRC staff requested that GPC provide a detailed discussion of the applicability of the Final Policy Statement criteria to each of the existing VEGP TS requirements proposed for relocation as well as the document and NRC regulation that will control each relocated requirement. The requested information is provided in the enclosure.

Sincerely, 7 7. , , C. K. McCoy

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. Enclosure l xc: Georcia Power Comoany i

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I Mr. J. B. Beasley, Jr.

Mr. M. Sheibani NORMS U. S. Nuclear Regulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. L. L. Wheeler, Licensing Project Manager, NRR Mr. C. R. Ogle, Senior Resident Inspector, Vogtle i

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS i CRITERIA APPLICATION REPORT TABLE 3 )

I EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

l. SPECIFICATION 3.1.1.1, Shutdown Margin (SDM) - Modes 1 and 2 I

11. SYSTEM / PARAMETER DESCRIPTION SDM is an initial condition in the safety analysis for such events as the zero power steamline break and rod ejection from subcritical condition.

Additionally, the analyses for various plant transients assume a successful reactor trip for accident mitigation, which in turn depends on sufficient SDM being present. The requirements for SDM in Modes 3,4, and 5 are addressed by the retained specification for SDM in those Modes.

I Ill. POLICY STATEMENT CRITERIA EVALUATION SDM is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant i pressure boundary. Therefore, SDM does not satisfy criterion 1. -

SDM is a process variable that is an initial condition of a DBA or transient .

analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, SDM is not a process variable  ;

that can be monitored by instrumentation in the control room and directly ,

controlled by the operator. During Modes 1 and 2 the SDM available from a .

reactor trip is determined by the inherent reactivity worth of the control rods.

SDM is verified by observation that the rods are above the insertion limits.  :

To ensure adequate SDM exists in Modes 1 and 2, the required rod insertion ,

limits are specified in the retained LCOs for Control and Shutdown Bank i insertion Limits. Therefore, although SDM satisfies criterion 2, a separate i Technical Specification for SDM is not necessary for Modes 1 and 2.

SDM is not a structure, system, or component that is part of the primary .

success path and which functions or actuates to mitigate a DBA or Transient i that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, SDM does not satisfy Criterion 3.  ;

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I VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS '

CRITERIA APPLICATION REPORT- TABLE 3 ~

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS ,

Ill. POLICY STATEMENT CRITERIA EVALUATION (continued)

SDM is not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, SDM does not satisfy Criterion 4.

IV. DISPOSITION  !

Consistent with NUREG-1431, SDM surveillance requirements that apply to ,

control rods are retained in either the Rod Group Alignment Limits or Control Bank Insertion Limits specification. Also consistent with NUREG-1431, the core reactivity balance surveillance is retained in the new LCO for Ccre Reactivity. The remaining Technical Specification requirements for GDM in Modes 1 and 2 will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATIONS 3.1.2.1, Flow Path - Shutdown 3.1.2.2, Flow Paths - Operating 3.1.2.3, Charging Pump - Shutdown 3.1.2.4, Charging Pumps - Operating 3.1.2.5, Borated Water Sources - Shutdown 3.1.2.6, Borated Water Sources - Operating II. SYSTEM / PARAMETER DESCRIPTION These Technical Specifications have been addressed as a group because they are all part of the boration subsystem, have common functional requirements, and the same relationship to DBAs.

The boration subsystem of the Chemical and Volume Control System (CVCS) provides the means to meet one of the functional requirements of the CVCS, i.e. to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the SDM. To accomplish this functional requirement the current specifications require a source of borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head.

The boration subsystem is not assumed to be operable to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCF,, which causes a boron dilution event, the response required by the on'"ator, is to close the appropriate valves in the reactor makeup system.

This action is required before the shutdown margin is lost. Operation of the boration subsystem equipment / flow paths specified in the technical specifications listed above is not assumed to mitigate this event. .

Operability of the boration subsystem components that are required to l mitigate a DBA or transient are addressed in the Technical Specifications for the ECCS (chapter 3.5).

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS Ill. POLICY STATEMENT CRITERIA EVALUATION The specifications for the boration subsystem do not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, the boration subsystem does not satisfy criterion 1.

The specifications for the boration subsystem do not contain requirements for a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the boration subsystem does not satisfy criterion 2.

The specifications for the boration subsystem do not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the boration subsystem does not satisfy Criterion 3.

The specifications for the boration subsystem do not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, the boration subsystem does not satisfy Criterion 4.

IV. DISPOSITION The Technical Specification requirements listed above for the boration subsystem will be relocated to the Technical Requirements Manual (TRM).

Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.1.3.3, Position Indicating Systems - Shutdown 3.10.5, Position Indicating Systems - Shutdown (Test Exception)

II. SYSTEM / PARAMETER DESCRIPTION The control rod position indicating system provides indication of rod position to the operator. This indication is used by the operator to verify that the rods are correctly positioned. In operating Modes, this indication is used during reactor startup to monitor rod position and to verify that the rods are inserted into the core immediately following a reactor trip. Rod position indication requirements during operation (Modes 1 and 2) are addressed by the retained LCO, " Rod Position Indication".

During the shutdown Modes the position indicating system provides information only and is not relied on by the operators to assist in the mitigation of a DBA or transient.

111. POLICY STATEMENT CRITERIA EVALUATION The specifications for Position Indicating Systems - Shutdown do not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Position Indicating Systems - Shutdown does not satisfy criterion 1.

l The specifications for Position Indicating Systems - Shutdown do not contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Position Indicating Systems - Shutdown does not satisfy criterion 2.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 I

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS- i l

111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specifications for Position Indicating Systems - Shutdown do not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Position Indicating Systems -

Shutdown does not satisfy Criterion 3.

The specifications for Position Indicating Systems - Shutdown do not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Position Indicating Systems - Shutdown does not satisfy Criterion 4.

IV. DISPOSITION f

The LCO for Position Indicating Systems - Shutdown will be relocated to the Technical Requirements Manual (TRM) along with the associated test exception LCO. The Position Indicating Systems - Shutdown surveillance requirement will be retained in the LCO for Position Indicating Systems, +

Modes 1 and 2. Retention of the surveillance requirement is consistent with the NRC guidance in Reference 2. Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS i CRITERIA APPLICATION REPORT ' TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l I

1. SPECIFICATION  !

3.1.3.4, Rod Drop Time

11. SYSTEM / PARAMETER DESCRIPTION This specification ensures that the safety analysis assumption for control rod j insertion time is preserved.The intended safety function of the control rods '

(reactor trip) is to place the reactor in a subcritical condition when a trip -

setpoint is exceeded.

The negative reactivity insertion following a reactor trip is a function of the position of the Rod Control Cluster Assemblies (RCCAs), and the variation in i rod worth as a function of rod position. The critical parameter as used in the safety analysis, is the time of insertion up to the dashpot entry. .,

The surveillance to verify rod drop time is retained in the LCO for Rod Group .t Alignment Limits as part of the control rod operability requirements. The retention of this surveillance is consistent with the NRC guidance provided in Reference 2. As such, the verification of the safety analysis assumption for f control rod insertion time remains in the Technical Specifications.

Ill. POLICY STATEMENT CRITERIA EVALUATION  ;

The specification for Rod Drop Time does not contain requirements for l installed instrumentation that is used to detect, and indicato in the control i room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Rod Drop Time does not satisfy criterion 1. -

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 ,

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFIC ATIONS lil. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specification for Rod Drop Time does contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the surveillance which verifies the safety analysis assumptions regarding rod drop time is retained in the technical specification for Rod Group Alignment Limits as part of control rod operability. Therefore, the remaining requirements of the Rod Drop Time specification proposed for relocation do not satisfy criterion 2.

The specification for Rod Drop Time does not contain requirements for a -

structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Rod Drop Time does not satisfy Criterion 3.

The specification for Rod Drop Time does not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Position Indicating Systems - Shutdown does not satisfy Criterion 4.

IV. DISPOSITION The LCO for Rod Drop Time will be deleted. The Rod Drop Time surveillance requirement that verifies the value of the safety analysis assumption associated with this specification is being retained in the Rod Group Alignment Limits specification which has the same Mode of Applicability as the Rod Drop Time specification. The retained surveillance becomes a l requirement for control rod operability. As such, the retained surveillance  !

and associated LCO effectively replace the function of the Rod Drop Time LCO. Retention of the Rod Drop Time surveillance requirement is consistent with the NRC guidance in Reference 2. l l

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l VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 i

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.3.3.2, Movable Incore Detectors

11. SYSTEM / PARAMETER DESCRIPTION This specification ensures the operability of the Movable Incore Detector Instrumentation when required to monitor the flux distribution within the core. The Movable incore Detector System is used for periodic surveillance of the power distribution, and calibration of the excore detectors. This surveillance verifies that the peaking factors are within their design envelopa.

Other (retained) Technical Specifications address the required power distribution limits, peaking factors, and excore detector calibrations, as well as their required surveillances. The movable incore detector instrumentation provides information only and is not considered in any DBA or transient analyses.

Ill. POLICY STATEMENT CRITERIA EVALUATION The specification for Movable incore Detectors does not contain requirements for installed instrumentation that is used to detect, and indicate '

in the control room, a significant abnormal degradation of the reactor coolant.

pressure boundary. Therefore, Movable incore Detectors do not satisfy criterion 1.

The specification for Movable incore Detectors does not contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the Movable Incore Detectors do not satisfy criterion 2.

The specification for Movable incore Detectors does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the Movable incore Detectors do not satisfy Criterion 3.

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b VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3  ;

1 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS' Ill. ' POLICY STATEMENT CRITERIA EVALUATION (continued) i The specification for Movable incore Detectors does not contain ,

requirements for a structure, system, or component which' operating 3 experience or probabilistic safety assessment (Reference li has shown to be _ i significant to public health and safety. Therefore, the Movable incore Detectors do not satisfy Criterion 4.

IV. DISPOSITION  :

The specification for Movable Incore Detectors will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be i controlled consistent with the provisions of 10 CFR 50.59.

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a VOGTLE UNITS 1 AND 2 TECHNICAL SFECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.3.3.3, Seismic Instrumentation ll. SYSTEM / PARAMETER DESCRIPTION Seismic Instrumentation is used to record data for use in evaluating the effect of a seismic event after the occurrence of such an event. The Seismic Instrumentation is not relied upon by operators to take immediate action in the event of an earthquake. Additionally, the Seismic Instrumentation is not used to mitigate a DBA or transient or assumed to function in any safety analysis.

Ill. POLICY STATEMENT CRITERIA EVALUATION The specification for Seismic Instrumentation does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Seismic Instrumentation does not satisfy criterion 1.

The specification for Seismic Instrumentation does not contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Seismic Instrumentation does not satisfy criterion 2.

The specification for Seismic Instrumentation coes not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitiga"e a DBA~or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Seismic Instrumentation does not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 ' ,

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS  ;

lli. POLICY STATEMENT CRITERIA EVALUATION (continued) ,

The specification for Seismic Instrumentation does not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. . Therefore, Seismic Instrumentation does not satisfy Criterion 4.

IV. DISPOSITION The specification for Seismic Instrumentation will be relocated to the Technical Requirements Manual (TRM). Changes to the THM will be controlled consistent with the provisions of 10 CFR 50.59.

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1 L VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 ,

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS-il. SPECIFICATION t 3.3.3.4, Meteorological Instrumentation ll. SYSTEM / PARAMETER DESCRIPTION -

Meteorological Instrumentation is used to record meteorological data for use in evaluating the effect of an accidental radioactive release from the plant.

The Meteorological Instrumentation provides information only and is not used to mitigate a DBA or transient or assumed to function in any safety analysis.

Ill. POLICY STATEMENT CRITERIA EVALUATION The specification for Meteorological Instrumentation does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant '

pressure boundary. Therefore, Meteorological Instrumentation does not satisfy criterion 1.

The specification for Meteorological Instrumentation does not contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes -

the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Meteorological Instrumentation does not satisfy criterion 2.

The specification for Meteorological Instrumentation does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Meteorological Instrumentation does not satisfy Criterion 3.

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EVALUATION AND DISPOSITION OF Ri! LOCATED TECHNICAL SPECIFICATIONS J lli. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specification for Meteorological Instrumentation does not contain requirements for a structure, system, or component which operating -

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experience or probabilistic safety assessment (Reference 1) has shown to be ,

-significant to public health and safety. Therefore, Meteorological Instrumentation does not satisfy Criterion 4.

IV. DISPOSITION  :

The specification for Meteorological Instrumerstation will be relocated to the .

Technical Requirements Manual (TRM). Changes to the TRM will be -l controlled consistent with the provisions of 10 CFR 50.59. j l

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~ VOGTLE UNITS 1" AND 2 TECHNICAL SPECIFICATIONS '-

CRITERIA APPLICATION REPORT TABLE 3- ,

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPEC'lFICATIONS '

l. SPECIFICATION 3.3.3.11, High Energy Line Break isolation Sensors (Vogtle specific TS)

II. -SYSTEM / PARAMETER DESCRIPTION See LCO 3.3.3.11, Section 111 Discussion in Table 2 of this report.

Ill. POLICY STATEMENT CRITERIA EVALUATION '

See LCO 3.3.3.11, Section 111 Discussion in Table 2 of this report.

IV. DISPOSITION l The specification for High Energy Line Break isolation Sensors will be -

relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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e VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 i

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.3.4, Turbine Overspeed Protection ll. SYSTEM / PARAMETER DESCRIPTION The Turbina Overspeed Protection instrumentation trips the turbine to prevent the generation of potentially damaging missiles from the turbine, in the event of a loss of the turbine speed control system, or a transient.

However, the turbine overspeed event is not a DBA and the Turbine Overspeed Protection Instrumentation is not considered in any DBA or transient analyses. A separate license amendment request to relocate this TS was submitted prior to the VEGP Improved TS conversion request and has been subsequently approved by the NRC (TAC numbers M88395 and M88396). Amendment numbers 88 and 66 were issued July 3,1995 to remove this TS from the VEGP Unit 1 and Unit 2 operating licenses.

Ill. POLICY STATEMENT CRITERIA EVALUATION The specification for Turbine Overspeed Protection Instrumentation does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Turbine Overspeed Protection Instrumentation does not satisfy criterion 1.

The specification for Turbine Overspeed Protection Instrumentation does not contain requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Turbine Overspeed Protection Instrumentation does not satisfy criterion 2.

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L VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specification for Turbine Overspeed Protection Instrumentation does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Turbine Overspeed Protection Instrumentation does not satisfy Criterion 3.

The specification for Turbine Overspeed Protection Instrumentation does not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Turbine Overspeed Protection Instrumentation does not satisfy Criterion 4.

IV. DISPOSITION The specification for Turbine Overspeed Protection Instrumentation has been relocated from the TS in accordance with License Amendments 88 and 66 issued by the NRC July 31995. The contents of the Turbine Overspeed Protection Instrumentation TS will be controlled in the FSAR in accordance with 10 CFR 50.59, as stated in the License Amendment Request originally made by GPC, or will be controlled in the Technical Requirements Manual in accordance with 10 CFR 50.59.

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EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l i L i

1. SPECIFICATION 3.4.2.1, Safety Valves - Shutdown

11. SYSTEM / PARAMETER DESCRIPTION The pressurizer safety valves protect the RCS from being pressurized above the RCS pressure Safety Limit. The pressurizer safety valves provide overpressurization protection during both power operation and hot standby.

However, the pressurizer safety valve function is not assumed in any safety analysis for the mitigation of a DBA or transient in Modes 4 and 5.

Ill. POLICY STATEMENT CRITERIA EVALUATION i The specification for Safety Valves - Shutdown does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Safety Valves - Shutdown does not satisfy criterion 1.

The specification for Safety Valves - Shutdown does not contain i requirements for process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Safety Valves - Shutdown does not satisfy criterion 2.

The specification for Safety Valves - Shutdown does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product carrier. Therefore, Safety Valves - Shutdown does not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3  ;

I EVALUATION AND DISPOSITl3N OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specification for Safety Valves - Shutdown does not contain requirements for a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Safety Valves - Shutdown does not satisfy Criterion 4. ,

IV. DISPOSITION

. Existing VEGP TS 3/4.4.2.1, " Reactor Coolant System - Safety Valves -

Shutdown," although identified for relocation, may instead be deleted based on the following discussion. NUREG-1431 requires the pressurizer safety valves to be operable in Modes 1,2, and 3, and Mode 4 with all RCS cold leg temperatures greater than the low temperature overpressure protection system enable temperature. At VEGP, the enable temperature is 350 *F, which is the Mode 3/ Mode 4 transition temperature. At VEGP, the necessary overpressure protection below 350 *F is provided by either the PORVs, the RHR suction relief valves, or a combination of the two in accordance with the Cold Overpressure Protection System, LCO 3.4.12. In addition, the setpoint for the code safeties (2485 i 1%) is well above that  ;

which is necessary to protect the RCS during low temperature operation.

Therefore, consistent with NUREG-1431, there is no need to require one pressurizer code safety to be operable in Modes 4 and 5.

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. VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCA FED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.4.5, Steam Generators ll. SYSTEM / PARAMETER DESCRIPTION This specification ;;rovides the inspection requirements for the Steam I Generator tubes, to ensure their structural integrity. The Steam Generators function to mitigate DBAs or transients by their heat removal capability.

However, as stated above, Specification 3.4.5 only addresses Steam Generator tube integrity based on the inspection requirements in the surveillance. Operability requirements regarding the heat removal function of I the Steam Generators are addressed in the (retained) RCS loop j specifications. In addition, consistent with the NRC guidance of Reference 2, the Steam Generator Tube Inspection requirements associated with this l specification have been retained in the Technical Specifications (Administrativo Controls Program) and are required to be met in the RCS l Operational Leakage specification surveillances. Therefore, the requirements of this specification are effectively retained within the RCS Operational Leakage specification and the Administrative Controls Program for Steam Generator Tube Inspection. I lil. POLICY STATEMENT CRITERIA EVALUATION The specification for Steam Generator tube inspections does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, the Steam Generator tube inspections do not satisfy criterion 1.

The specification for Steam Generator tube inspections does not contain requirements for process variables, design features, or operating restrictions .

that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the Steam Generator tube inspections do not satisfy criterion 2.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3  !

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l 1

l 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The specification for Steam Generator tube inspections does contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, specific operability requirements regarding the required heat removal function and tube leakage limits of the Steam Generators are addressed in the (retained) RCS Loop and Operational Leakage specifications. Therefore, although Steam Generators satisfy criterion 3, a separate LCO for Steam Generator tube inspection requirements is not necessary.

The Steam Generator tube inspections are not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety.

Therefore, Steam Generator tube inspection surveillances do not satisfy Criterion 4.

IV. DISPOSITION Consistent with the NRC guidance in Reference 2, the Steam Generator Tube inspection requirements associated with LCO 3.4.5 have been retained in the ,

Technical Specifications (Administrative Controls Program) and are required to be met in the RCS Operational Leakage specification surveillances.

Additionally, the RCS Operational Leakage specification operability requirements have the same Mode of Applicability as LCO 3.4.5. Since the requirements of LCO 3.4.5 are effectively retained within the RCS Operational Leakage specification and the Administrative Controls Program for Steam Generator Tube Inspection this LCO may be deleted instead of relocated.

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.VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.4.7, RCS Chemistry i

11. SYSTEM / PARAMETER DESCRIPTION This specification places limits on the oxygen, chloride and fluoride content L in the RCS to minimize corrosion. Minimizing corrosion of the RCS will i reduce the potential for RCS leakage or failure due to stress corrosion, and ultimately ensure the structuralintegrity of the RCS. However, degradation '

of the RCS boundary is a long term process for which other more direct .

methods of monitoring are available. The inservice inspections required by regulations (10 CFR 50.55a) and the RCS leakage limits in Technical .

Specifications are examples of requirements provided to monitor and prevent long-term degradation of the RCS boundary materials. In addition, these requirements provide for long-term maintenance of acceptable RCS conditions.

Ill. POLICY STATEMENT CRITERIA EVALUAT10N ,

The RCS Chemistry is not installed instrumentation that is used to detect, i and indicate in the control room, a significant abnormal degradation of the

. reactor coolant pressure boundary. Therefore, RCS Chemistry does not satisfy criterion 1.

The RCS Chemistry is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, RCG Chemistry does not satisfy criterion 2.

The RCS Chemistry is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, RCS chemistry does not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The RCS Chemistry is not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, RCS Chemistry does not satisfy Criterion 4.

IV. DISPOSITION The specification for RCS Chernistry will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS  !

CRITERIA APPLICATION REPORT TABLE 3 l EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.4.9.2, Pressure / Temperature Limits, Pressurizer

11. SYSTEM / PARAMETER DESCRIPTION Since the pressurizer normally operates in temperature ranges above those for which there is a reason for concern of nonductile failure, pressure and temperature limits are placed on the pressurizer to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. However, a failure of pressurizer integrity would result in an analyzed event (loss of coolant accident) for which numerous systems and components are required and retained in the Technical Specifications.

Therefore, the pressurizer pressure / temperature limits are not relied on to prevent or mitigate a DBA or transient, nor are these limits an operating restriction that is required to preclude an unanalyzed accident or transient j (criterion 2).

Ill. POLICY STATEMENT CRITERIA EVALUATION The Pressurizer Pressure / Temperature Limits are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Pressurizer Pressure / Temperature Limits do not satisfy criterion 1.

The Pressurizer Pressure / Temperature Limits are not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to l

the integrity of a fission product barrier. Therefore, Pressurizer Pressure / Temperature Limits do not satisfy criterion 2.

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The Pressurizer Pressure / Temperature Limits are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the f ailure of or j presents a challenge to the integrity of a fission product barrier. Therefore, Pressurizer Pressure / Temperature Limits do not satisfy Criterion 3.

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1 VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA-APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

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Ill. POLICY STATEMENT CRITERIA EVALUATION (continued) f The Pressurizer Pressure / Temperature Limits are not a structure, system, or

. component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. .

Therefore, Pressurizer Pressure / Temperature Limits do not satisfy Criterion 4. l IV. DISPOSITION The specification for Pressurizer Pressure / Temperature Limits will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM -

will be controlled consistent with the provisions of 10 CFR 50.59.

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I-l VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS i

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1. SPECIFICATION 3.4.10, RCS Structural Integrity ll. SYSTEM / PARAMETER DESCRIPTION This specification refers to the inservice inspection and test requirements for {

the ASME Code Class 1,2 and 3 components. The requirements referred to by this specification prevent long-term degradation of ASME Code Class 1, 2, and 3 components. In addition to the ASME inspection and test requirements, this specification includes surveillance requirements for inspection of the reactor coolant pump flywheel and main steam and feedwater lines.

Ill. POLICY STATEMENT CRITERIA EVALUATION l This specification does not contain requirements for installed instrumentation l

that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, the RCS Structural Integrity specification does not satisfy criterion 1.

i This specification does not contain requirements for a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the RCS Structural Integrity specification does not satisfy criterion 2.

ASME Code Class 1,2 and 3 components are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of or present a challenge to the integrity of a fission product barrier.

Individual ASME Code Class 1,2 and 3 components may satisfy criterion 3 and the requirements that ensure the integrity / operability of these components are included in the individual specifications that cover these components. Additionally, the test and inspection requirements referred to in the RCS Structural Integrity specification are all either retained in Technical Specifications or required by regulation. Therefore, the RCS Structural Integrity specification does not satisfy criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPOP.T- TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS lll POLICY STATEMENT CRITERIA EVALUATION (continued)

The requirements of RCS Structural Integrity specification are not a  :

structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health '!

and safety. Therefore, the RCS Structural Integrity specification does not satisfy Criterion 4.

IV. DISPOSITION Consistent with the NRC guidance in Reference 2 and NUREG-1431, the surveillance requirements associated with the RCS Structural Integrity specification are retained in Technical Specifications or regulations. The performance of inservice inspections in accordance with ASME Section XI is required by 10 CFR 50.55a and the requirement to perform inservice testing has been retained in the Administrative Controls section of the Technical

' Specifications as a Program. Additionally, the RCS Structural Integrity specification requirements for the inspection of the reactor coolant pump flywheel and main steam and feedwater lines have also been retained in  !

Administrative Control Programs in the Technical Specifications. These retained requirements assure the integrity and operability of the individual ASME Code Class 1,2 and 3 components that satisfy criterion 3. However, the appropriate Applicability, Actions, and Completion Times that apply to those individual components are included in the retained Technical  :

Specifications that address those components. The RCS Structural integrity specification is redundant to the applicable inspection and test requirements described above and the existing individual Technical Specifications which contain the operability requirements for the required components or '

equipment that meet criterion 3. Therefore, the RCS Structural Integrity specification may be deleted instead of relocated. I i

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1. SPECIFICATION 3.4.11, RCS Vents ll. SYSTEM / PARAMETER DESCRIPTION The RCS Vents are provided to exhaust noncondensible gases and/or steam from the RCS which could inhibit natural circulation core cooling following any event involving a loss of offsite power and requiring long term cooling, such as a loss-of-coolant accident (LOCA). Their f. unction, capabilities, and testing requirements are consistent with the requirements of item II.B.I of NUREG-0737, " Clarification of TMI Action Plan Requirements," however, the operation of RCS Vents is not assumed in any safety analysis. This is because the operation of the vents is not part of the primary success path. f The operation of these vents is an operator action after the event has occurred, and is only required when there is indication that natural circulation is not occurring.

Ill. POLICY STATEMENT CRITERIA EVALUATION -

The RCS Vents are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, RCS Vents do not satisfy criterion 1.

The RCS Vents are not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, RCS Vents do not satisfy criterion 2.

t The RCS Vents are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the i 1

integrity of a fission product barrier. Therefore, RCS Vents do not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS Ill. POLICY STATEMENT CRITERIA EVALUATION (continued)

The RCS Vents are not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, RCS Vents do not satisfy Criterion 4.

IV. DISPOSITION The specification for RCS Vents will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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1. SPECIFICATION 3.6.1.2, Containment Leakage ll. SYSTEM / PARAMETER DESCRIPTION This specification identifies the allowable leakage rates for the containment structure which are established in 10 CFR Part 50 APPENDIX J. ' Adherence to these limits ensures that the consequences of a DBA will not exceed the 10 CFR Part 100 limits.

111. POLICY STATEMENT CRITERIA EVALUATION 4

Containment Leakage is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Cor.tainment Leakage does not satisfy criterion 1.

Containment Leakage is an operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, consistent k with NUREG-1431, the requirement to meet the leakage limits established in 10 CFR Part 50 APPENDIX J has been retained in the " Containment" specification surveillances. Therefore, a'though containment leakage satisfies criterion 2, a separate Coni:ainment Leakage specification is no longer necessary.

Containment Leakage is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Containment Leakage does not satisfy Criterion 3.

Containment Leakage is not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Containment -

Leakage does not satisfy Criterion 4.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA' APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 4

IV. DISPOSITION Consistent with NUREG-1431, the requirement to meet the leakage limits established by 10 CFR Part 50 APPENDIX J are retained in the Containment specification. Therefore, the requirements of the Containment Leakage  ;

specification have been effectively retained in the Containment specification.  !

l The Containment specification Applicability, Actions and Completion Times .

I adequately address the leakage requirements. As such, the Containment Leakage LCO is redundant to the Containment LCO and therefore may be deleted instead of relocated.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.6.1.6, Containment Structural Integrity

' ll. SYSTEM / PARAMETER DESCRIPTION The containment serves as a barrier to prevent the release of fission products following a LOCA or steamline break inside containment. To mitigate the potential consequences of a DBA, it is necessary that the containment structure meet its functional requirements. Therefore, this specification requires that the capability of the containment structure to withstand peak accident pressure be demonstrated periodically. This specification outlines an appropriate containment tendon testing program which demonstrates this capability.

Ill. POLICY STATEMENT CRITERIA EVALUATION Containment Structural Integrity is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, Containment Structural Integrity does not satisfy criterion 1. j Containment Structural Integrity is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Containment Structural integrity does not satisfy criterion 2. l t

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION HEPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The Containment Structural Integrity specification contains requirements for a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, consistent with the NRC guidance of reference 2, the applicable requirements have been retained in the Technical Specifications as an Administrative Controls Program for Tendon Surveillance Testing. This program is required for containment operability in the surveillances of the Containment specification. Therefore, a separate Containment Structural Integrity specification is no longer necessary.

The Containment Structural Integrity testing requirements are not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, The Containment Structural Integrity testing requirements do not satisfy Criterion 4.

IV. DISPOSITION Consistent with the NRC guidance in Reference 2, the requirement for tendon surveillance testing is retained in an Administrative Controls Program that is a requirement for containment operability. The specific details of the required testing are relocated to the body of the Administrative Controls Program outside of the technical specifications. Changes to the Program will be controlled consistent with the provisions of 10 CFR 50.59. Therefore, the requirements of the Containment Structural Integrity specification have been effectively retained in the " Pre-Stressed Concrete Containment Tendon Surveillance Program" and the Containment specification. The Containment specification Applicability, Actions and Completion Times adequately address the Containment Structural Integrity requirements. As such, the Containment Structural Integrity LCO is redundant to the Containment LCO and Tendon Surveillance Program and therefore may be deleted instead of

. relocated. .

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS j

l. SPECIFICATION 1

3.7.2, Steam Generator Pressure / Temperature Limitation

11. SYSTEM / PARAMETER DESCRIPTION This specification places limits on the steam generator pressure and temperature to ensure that the pressure induced stresses are within the maximum allowable fracture toughness stress limits. The pressure and temperature limits are based on a steam generator RTuor sufficient to prevent brittle fracture. However, the failure of steam generator integrity results in analyzed events (steam generator tube rupture or other loss of coolant accident) for which numerous systems and components are required and retained in the Technical Specifications. Therefore, the Steam Generator Pressure / Temperature Limitation is not relied on to prevent or mitigate a DBA or transient, nor is this limitation an operating restriction that is required to preclude an unanalyzed accident or transient (criterion 2).

111. POLICY STATEMENT CRITERIA EVALUATION The Steam Generator Pressure / Temperature Limitation is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Therefore, the Steam Generator Pressure / Temperature Limitation does not satisfy criterion 1.

The Steam Generator Pressure / Temperature Limitation is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a ,

challenge to the integrity of a fission product barrier. Therefore, the Steam Generator Pressure / Temperature Limitation does not satisfy criterion 2.

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CRITERIA' APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS f Ill. POLICY STATEMENT CRITERIA EVALUATION (continued)

- The Steam Generator Pressure / Temperature Limitation is not a structure, system, or component that is part of the primary success path and which l functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,

Therefore, the Steam Generator Pressure / Temperature Limitation does not satisfy Criterion 3. 1 The Steam Genarator Pressure / Temperature Limitation is not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, the Steam Generator Pressure / Temperature Limitation does not satisfy Criterion 4.

IV. DISPOSITION The Steam Generator Pressure / Temperature Limitation specification will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3  ;

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EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS -l t

1. SPECIFICATION 1 3.7.8, Snubbers l

11. SYSTEM / PARAMETER DESCRIPTION The Snubbers prevent unrestrained pipe motion under dynamic loads. >

,. Snubbers also allow normal thermal expansion of piping and nozzles to  !

eliminate excessive thermal stresses during heatup or cooldown. Snubbers l are support systems for the primary components whose operation or function may be an assumption of a safety analysis, however, the function of a Snubber is not specifically assumed in any safety analysis. As support systems, Snubbers are required for the operability of the equipment or components they support, but are not considered part of the " primary success path" (criterion 3). The snubber inspection requirements of this specification are part of the inservice inspection program and are consistent i with the requirements of 10 CFR 50.55a.

• lli. POLICY STATEMENT CRITERIA EVALUATION t i

Snubbers are not installed instrumentation that is used to detect, and

• indicate in the control room, a significant abnormal degradation of the reactor i coolant pressure boundary. Therefore, Snubbers do not satisfy criterion 1.

Snubbers are not a process variable, design feature, or operating restriction ,

that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product i barrier. Therefore, Snubbers do not satisfy criterion 2. [

Snubbers are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the  ;

integrity of a fission product barrier. Therefore, Snubbers do not satisfy Criterion 3. 1 i

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

Snubbers (except in the roll as support equipment for components / equipment which do meet criterion 3 above) are not a structure, system, or component which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Snubbers do not satisfy Criterion 4.

IV. DISPOSITION The specification for Snubbers will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59 and 10 CFR 50.55a, including the approved exemptions. I 6

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.7.9, Sealed Source Contamination

11. SYSTEM / PARAMETER DESCRIPTION This specification ensures that leakage from Byproduct, Source and Special Nuclear Material sources will not exceed allowable intake values. The limitations on removable contamination for sources requiring leak testing, including c!pha emitters, is based on 10 CFR Part 70.39(a)(3) limits for plutonium. The requirements of this specification do not impact reactor operation or safety.

Ill. POLICY STATEMENT CRITERIA EVALUATION Sealed Source Contamination is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation f

of the reactor coolant pressure boundary. Therefore, Sealed Source Contamination does not satisfy criterion 1.

Sealed Source Contamination is not a process variable, design feature, or 4 operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Sealed Source Contamination does not satisfy criterion 2.

Sealed Source Contamination is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, Sealed Source Contamination does not satisfy Criterion 3.

Sealed Source Contamination is not a structure,' system, or component which operating expertence or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, Sealed Source Contamination does not satisfy Criterion 4.

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EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS IV. DISPOSITION

' The specification for Sealed Source Contamination will be relocated to the ,

Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10_ CFR 50.59.

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. VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT. TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l .' SPECIFICATION 3.7.12, Reactor Coolant Pump Thermal Barrier Cooling Water isolation l

lI. SYSTEM / PARAMETER DESCRIPTION See LCO 3.7.12, Section lli Discussion in Table' 2 of this report.

Ill. ' POLICY STATEMENT CRITERIA EVALUATION See' LCO 3.7.12, Section lli Discussion in Table 2 of this report.

IV. DISPOSITION The specification'for Reactor Coolant Pump Thermal Barrie'r Cooling Water Isolation will be relocated to the Technical Requirements Manual (TRM).

Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l l

l. SPECIFICATION 3.7.13, Diesel Generator Building and Auxiliary Feedwater Pumohouse ESF HVAC System il. SYSTEM / PARAMETER DESCRIPTION See LCO 3.7.13, Section ill Discussion in Table 2 of this report.

Ill. POLICY STATEMENT CRITERIA EVALUATION See LCO 3.7.13, Section ill Discussion in Table 2 of this report.

IV DISPOSITION The specification for Diesel Generator Building and Auxiliary Feedwater Pumphouse ESF HVAC System will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled  ;

consistent with the provisions of 10 CFR 50.59.

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l VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.8.4.1, Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers to Isolation Transformers Between 480 V Class 1E Busses and Non-class 1E Equipment ll. SYSTEM / PARAMETER DESCRIPTION The containment penetration conductor overcurrent protective devices and Feeder Breakers are installed to minimize the potential for a fault in a component inside containment, or in cabling which penetrates containment.

Containment electrical penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the operability of overcurrent circuit breakers. The l deenergizing of A.C. circuits inside containment minimizes the potential for a fault in a component inside containment from propagating to equipment outside containment and potentially damaging the penetration. However, containment penetration degradation is monitored and identified by the required 10 CFR 50 APPENDIX J Ieak rate testing. The operation of these _

overcurrent devices and breakers to deenergize A.C. circuits is not required for normal operation or accident mitigation and is not an assumption of any safety analyses. Additionally, these AC circuits are separated from class 1E circuits, and their failure will not degrade any Class 1E circuits.

Ill. POLICY STATEMENT CRITERIA EVALUATION The Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor .

coolant pressure boundary. Therefore, this specification does not satisfy criterion 1.

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l VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 1 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS Ill. POLICY STATEMENT CRITERIA EVALUATION (continued)

The Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers are not process variables, design features, or operating restrictions that are an initial condition of a DBA or transient analysi: that either assumes the failure of or presents a challenge to the integrity of a ,

fission product barrier. Therefore, this specification does not satisfy criterion 2.

The Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers are not structures, systems, or components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the )

integrity of a fission product barrier. Therefore, thir specification does not l satisfy Criterion 3.

The Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers are not structures, systems, or components which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, this specification does not satisfy Criterion 4.

IV. DISPOSITION The specification for Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers to Isolation Transformers Between l 480 V Class 1E Busses and Non-class 1E Equipment will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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1. SPECIFICATION -

3.8.4.2, Safety-Related Motor-Operated Valves Thermal Overload Protection and Bypass Devices ll. SYSTEM / PARAMETER DESCRIPTION The motor-operated valve thermal overload protection bypass ensures that the thermal overload will not prevent a motor-operated valve from performing its intended safety function. These protection bypass devices support the operability of the valves which contain them. The operability of the valves is required in turn to support the operability of their associated system.

Therefore, the operability of the thermal overload protection bypass devices - .

is determined in accordance with the operability requirements of the f L Technical Specifications for those systems containing valves designed with such devices.

111.

l POLICY STATEMENT CRITERIA EVALUATION The motor-operated valves thermal overload protection bypass jevices are not installed instrumentation that is used to detect, and indicate in the I control room', a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, this specification does not satisfy criterion 1.

l The motor-operated valves thermal overload protection bypass devices are

j. not process variables, design features, or operating restrictions that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Therefore, this specification does not satisfy criterion 2.

The motor-operated valves thermal overload protection bypass devices are i-not structures, systems, or components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either l assumes the failure of or presents a challenge to the integrity of a fission F

product barrier. Therefore, this specification does not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 1

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS Ill. POLICY STATEMENT CRITERIA EVALUATION (continued)

The motor-operated valves thermal overload protection bypass devices are not structures, systems, or components which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, this specification does not satisfy Criterion 4.

IV. DISPOSITION The specification for Safety-Related Motor-Operated Valves Thermal Overload Protection and Bypass Devices will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59 Page 45

VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 1

1. SPECIFICATION 3.9.3, Decay Time

11. SYSTEM / PARAMETER DESCRIPTION i

This specification places a time limit on reactor subcriticality prior to the I movement of irradiated fuel assemblies in the reactor vessel. This ensures that sufficient time has elapsed for the radioactive decay of short-lived fission products and is consistent with the assumptions used in the safety analysis. However, the schedule restraints of the activities required prior to moving irradiated fuelin the reactor vessel after a shutdown prevents the time limit of this specification from being exceeded. The preparations for moving fuel include RCS cooldown, depressurization, boration, removal of the reactor vessel head and upper internals and flooding the reactor cavity to the required level. This specification is not included in NUREG-1431.

j 111. POLICY STATEMENT CRITERIA EVALUATION Decay Time is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, this specification does not satisfy criterion 1.

Decay Time is an operating restriction or process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the practical aspects of preparing to move irradiated fuel immediately after a plant shutdown, which also include restrictions (i.e., Refueling Cavity Water Level) that are required by the Technical Specifications, prevents the Decay Time limit from being exceeded. Therefore, although Decay Time meets criterion 2, the retention of the specification for this limit is not necessary.

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!- VOGTLE UNITS'1 AND 2 TECHNICAL SPECIFICATIONS 1 CRITERIA APPLICATION REPORT TABLE 3 ' j EVALUATION AND DISPOSITION OF RELOCATFD TECHNICAL SPE FICATIONS q

,; ll1. POLICY STATEMENT CRITERIA EVALUATION (continued) .,

Decay Time is not a structure, system, or component that is part of the l primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the  :

integrity of a fission product barrier. Therefore, this specification does not i satisfy Criterion 3. l Decay Time is not a structure, system, or component which operating ,

experience or probabilistic safety assessment (Reference 's) has shown to be significant to public health and safety. Therefore, this specification does not satisfy Criterion 4.

t IV. DISPOSITION The specification for Decay Time will be relocated to the Technical '

Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59. -

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS

1. SPECIFICATION 3.9.5, Communications

11. SYSTEM / PARAMETER DESCRIPTION This specification requires communication between the control room and the l

refueling station, to ensure that any significant change in the facility status observed on the control room instrumentation can be communicated to the i refueling station personnel. However, this communication is not credited in any DBA or transient analyses.

Ill. POLICY STATEMENT CRITERIA EVALUATION The Communications specification does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Therefore, this specification does not satisfy criterion 1.

1 The Communications specification does not contain requirements for procc,as variables, design features, or operating restrictions that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy criterion 2.

7 The Communications specification does not contain requirements for l structures, systems, or components that are part of the primary success path l and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy Criterion 3.

The Communications specification does not contain requirements for structures, systems, or components which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, this specification does not satisfy l

Criterion 4.

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VOGTLE UNITS l' AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT' TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS IV. DISPOSITION The specification for Communications will be relocated to the Technical

-. Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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1. SPECIFICATION 3.9.6, Refueling Machine ll. SYSTEM / PARAMETER DESCRIPTION This specification ensures that the refueling machine and auxiliary hoist will have sufficient load capacity for their intended purpose and will be used ,

correctly during refueling. Additionally, this specification ensures that the i core internals and reactor vessel are protected from excessive lifting force during refueling operations. Although this specification contains requirements designed to prevent damage to fuel assemblies, core internals, and reactor vessel, these requirements are not relied upon to prevent or mitigate the consequences of a DBA.

Ill. POLICY STATEMENT CRITERIA EVALUATION The Refueling Machine specification does not contain requirements for-installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure ;

boundary. Therefore, this specification does not satisfy criterion 1.  %

l The Refueling Machine specification does not contain requirements for 'l process variables, design features, or operating restrictions that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, j this specification does not satisfy criterion 2.

The Refueling Machine specification does not contain requirements for structures, systems, or components that are part of the pn, y success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy Criterion 3.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS 111. POLICY STATEMENT CRITERIA EVALUATION (continued)

The Refueling Machine specification does not contain requirements for structures, systems, or components which operating experience or probabilistic safety assessment (Reference 1) has shown to be significant to public health and safety. Therefore, this specification does not satisfy Criterion 4.

IV. DISPOSITION The Refueling Machine specification will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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VOGTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REPORT TABLE 3

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS j

l. SPECIFICATION j

3.9.7, Crane Travel- Spent Fuel Storage Pool Building '

II. SYSTEM / PARAMETER DESCRIPTION This specification ensures that loads in excess of the nominal weight of one l

fuel assembly containing a Rod Control Cluster Assembly, plus the weight of l the fuel handling tool, will not be moved over other fuel assemblies stored in the spent fuel storage racks, in the event that the load is dropped, the activity released is limited to that contained in one fuel assembly. This also prevents any possible distortion of fuel assemblies in the storage racks from resulting in a critical configuration. Although this specification supports the maximum refueling accident assumption in the DBA, the fuel handling crane travellimits specified are not process variables monitored and controlled by the operator during operation (criterion 2). The limits in this specification are verified on a periodic basis and apply to the design (physical stops).

Ill. POLICY STATEMENT CRITERIA EVALUATION The Crane Travel- Spent Fuel Storage Pool Building specification does not contain requirements for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Therefore, this specification does not satisfy criterion 1.

The Crane Travel - Spent Fuel Storage Pool Building specification does not contain requirements for process variables, design features, or operating restrictions that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy criterion

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CRITERIA APPLICATION REPORT ' TABLE 3 EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS. ,

111. POLICY STATEMENT CRITERIA EVALUATION (continued) ,

The Crane Travel- Spent Fuel Storage Pool Building specification does not  !

contain requirements for structures, systems, or components that are part of  :

the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the  !

integrity of a fission product barrier. Therefore, this specification does not satisfy Criterion 3.

The Crane Travel- Spent Fuel Storage Pool Building specification does not contain requirements for structures, systems, or components which operating experience or probabilistic safety assessment (Reference 1) has i shown to be significant to public health and safety. Therefore, this specification does not satisfy Criterion 4. t IV. DISPOSITION The Crane Travel - Spent Fuel Storage Pool Building specification will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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- CRITERIA APPLICATION REPORT TABLE 3 l

EVALUATION AND DISPOSITION OF RELOCATED TECHNICAL SPECIFICATIONS l l .' SPECIFICATION . )

3.9.12, Fuel Handling Building Post Accident Ventilr. tion System II. SYSTEM / PARAMETER DESCRIPTION i

g See LCO 3.9.12, Section lli Discussion in Table 2 of this report. i l

Ill. POLICY STATEMENT CRITERIA EVALUATION l-See LCO 3.9.12, Section til Discussion in Table 2 of this report.

IV. DISPOSITION

~

The specification for the Fuel Handling Building Post Accident Ventilation System will be relocated to the Technical Requirements Manual (TRM).

Changes to the TRM will be controlled consistent with the provisions of 10 CFR 50.59.

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